1. Field of the Invention
The present invention relates to a process for recovery of uranium (U) from a spent nuclear fuel using a highly alkaline carbonate solution and, more particularly, to a process for recovery of uranium alone (i.e., in isolation) from a spent nuclear fuel using a highly alkaline carbonate solution, characterized by excellent proliferation resistance that prevents dissolution-leaching of transuranium element (“TRU”) nuclides such as plutonium (Pu), neptunium (Np), americium (Am), curium (Cm), etc., from the spent nuclear fuel, as well as providing ecologically-friendly (“eco-friendly”) characteristics that minimize waste generation.
2. Description of the Related Art
Recently, global warming due to accelerated fossil energy consumption and an increase in environmental hazards caused there from, as well as the depletion of natural resources (e.g., oil) have triggered intense international research and development in order to ensure access to a stable supply of environmentally clean energy. To date, research and proposals in the area of next-generation nuclear fuel cycle management have resulted in a generation-IV (“GEN-IV”) concept, which was introduced by the International Atomic Energy Agency (“IAEA”) in 2001; an Advanced Fuel Cycle Initiative (“AFCI”), which was introduced in 2003; as well as a Global Nuclear Energy Partnership (“GNEP”), which introduced in 2006 by the U.S. Department of Energy (“the DOE”). Such programs focus on nuclear sustainability, proliferation resistance, system stability, system reliability, and attempt to minimize the environmental impact of nuclear reactors by effective recycling of valuable elements in the spent nuclear fuel.
GEN-IV and GNEP focus on minimizing waste, proliferation, and environmental impact recycling or disposing spent nuclear fuel using a wet and/or a dry treatment process(es). A wet process typically utilizes one or more solvent extractions after dissolution of the spent nuclear fuel in a strong acid, which separate and treat the nuclides in the spent nuclear fuel according to different purposes. On the other hand, a dry process typically uses electrolytic treatment of the spent nuclear fuel in a molten salt.
The Plutonium-Uranium Extraction process (“PUREX”) is representative of wet process methods that can separate U and Pu with high purity. However, PUREX has a critical defect for proliferation resistance because excessive waste is generated by this wet process. Therefore, PUREX belongs to spent nuclear fuel treatment concepts that must be avoided in views of the advanced nuclear fuel cycle.
Recently, U.S. AFCI reports disclosed a Uranium Extraction Plus process (“UREX+”), which is an advanced wet process that will be applied in GNEP programs. In the UREX+ process, radioactive toxicities are removed from spent fuel to provide high-purity U having toxicity comparable to naturally occurring U that is separated from the spent nuclear fuel. Such low toxicity U can, in turn produce a low level (radioactive) waste and can allow for near-surface disposal and/or containment. In the UREX+ process, since Pu is not separated alone but is maintained in mixture with TRU and/or fission product (“FP”), the process is considered to have favorable proliferation resistance.
As for GNEP, a number of studies have attempted to separate a series of TRU nuclides based on their chemical properties via solvent extraction by applying advanced UREX+ II, III and/or IV programs, in order to reuse the spent nuclear fuel and minimize high level waste generation, thereby considerably improving ability of a site for radioactive waste disposal. A variety of wet processes studied first in the EU and Japan substantially adopt the solvent extraction with high separation efficiency in order to minimize waste generation and reuse (or recycle) useful nuclides contained in the spent nuclear fuel.
As one of the dry processes for spent nuclear fuel, a PYRO-metallurgy (“PYREX”) process is used to separate U, TRU, rare earths (RE), etc. from spent nuclear fuel in a LiCl or LiF molten salt at a temperature of 500° C. or more using a series of electrolysis steps. Such processes have been reported in Japan, Russia and the U.S. since the 1960s, and require a high technical level for high purity separation of nuclides and favorable proliferation resistance because Pu is mixed with TRU or others and reacts with the same. Drawbacks to the PYREX process are high levels of waste, a need for corrosion resistant materials required for use in a hot molten salt atmosphere, a lack of practical technologies concerning the same, and the like.
In order to overcome problems in current dry processes used in the U.S., Japan, EU, etc., such as operation stability, development of anti-corrosive materials, high purity separation, and the like, extensive long term research and investigation into the concerned techniques are required. The U.S. has proposed to commercialize a wet separation process by 2015, while a dry separation technique is planned become commercially available after 2030. That is, development of a wet process currently takes precedence over the development of an improved dry process. Therefore, current research and development in most countries has focused on the use of wet processes to treat spent nuclear fuel and high level waste generated therefrom rather than dry processes.
Compared to dry processes, most wet processes can operate at lower temperature and can be operated continuously. However, wet processes generally exhibit reduced proliferation resistance, for example, because most organic solvent extractants exhibit extraction properties only in acidic atmospheres (to enable separation of Pu ions using a solvent extractant), and it is difficult to selectively dissolve elements from spent nuclear fuel in a strong acid solution (which is used for most wet processes).
Generally, the solvent extraction used in the above wet process is technologically similar to the solvent extraction used in PUREX, which is an example of the wet process for Pu separation. However, PUREX or UREX+, which are common wet processes that utilize solvent extraction, are substantially different from the present invention in terms of technical characteristics and purposes to be achieved. Although U.S. Pat. Nos. 4,410,487, 4,436,704, 4,460,547, 4,675,166 and 4,696,768 disclose processes for U separation in a carbonate solution, purposes, procedures, and conditions of the medium used in these patents are substantially different from those of the present invention. U.S. Pat. No. 5,640,668 describes a process for separating actinide elements such as Pu and Np under alkaline conditions. U.S. Pat. Nos. 4,410,497 and 6,471,922 disclose a process for precipitating and separating U under alkaline conditions. U.S. Pat. No. 5,384,104 discloses a technique for recovering U from a radioactive waste using a carbonate solution. U.S. Pat. No. 4,305,911 proposes U recovery from U raw ores using an alkaline medium. Japanese Laid open Patent No. 1997-113681 and J. Nucl. Sci. & Tech. 43:255-262 (2006) disclose a process for treating a spent nuclear fuel using a carbonate system. However, none of the systems described in these various references provides a commercially reliable nuclear waste treatment system that can isolate U in an efficient manner.